This invention relates to a process for the separation and purification of curie quantities of yttrium-90, produced from strontium-90 for use in the treatment of cancer and rheumatoid arthritis of the knee joint. More specifically, the invention relates to a process for the separation and purification of yttrium-90 while preserving the purity of the strontium-90 source, and removing other impurities from the yttrium-90.
Considerable effort has recently been directed toward the development of site-specific methods for the treatment of various forms of cancer using radionuclides. The suitability of a given radionuclide for use in such an application is determined by several factors, among them its mode of production, the availability of effective methods for its attachment to a site-specific agent, its potential toxicity if detached from the agent, and its therapeutic effectiveness. These considerations, together with the requirement that the isotope have an appropriate half-life and decay scheme, severely limit the number of radionuclides which may be seriously considered for use in radioimmunotherapy.
Among the more attractive radionuclides for therapeutic applications is yttrium-90. Its relatively short half-life (64.0 h) and maximum beta energy (2.28 MeV) make it well suited for a variety of applications, ranging from the radiolabeling of antibodies for tumor therapy to the production of radiolabeled particles for the treatment of liver malignancies. Yttrium-90 results from the decay of strontium-90 according to the following scheme: ##EQU1##
Before it can be safely employed in clinical applications, the .sup.90 Y must be made essentially free of .sup.90 Sr, an isotrope known to cause bone marrow suppression. In addition, any trace elements which could interfere with the radiolabeling process by competing with .sup.90 Y for binding sites must be removed.
Many methods for effecting the separation of .sup.90 Y and .sup.90 Sr have been described, among them solvent extraction, ion-exchange, precipitation, and various forms of chromatography. Of these, ion-exchange methods have probably received the most attention. Numerous procedures have been reported, for example, in which a cation exchange resin (e.g. Dowex 50) is used to retain .sup.90 Sr, while the .sup.90 Y is eluted with an aqueous complexant such as lactate, acetate, citrate, oxalate, or EDTA. Several of these procedures have been proposed as the basis for .sup.90 Y generator systems. Unfortunately, in each case, the .sup.90 Y is not eluted in a form suitable for direct labeling of antibodies. That is, often the concentration of the complexing agent is such that it will compete effectively with binding sites on the antibody for the activity, resulting in a decrease in the labeling efficiency. Thus, it usually becomes necessary to either remove these materials prior to antibody labeling or to carry out postlabeling purification. In addition, ion-exchange resins are plagued by a gradual loss of capacity due to radiation damage. As a result, ion-exchange is considered suitable only for subcurie quantities of activity, less than the quantities which often must be processed for clinical applications. Finally, achieving acceptable .sup.90 Y yields while minimizing .sup.90 Sr breakthrough often requires the use of long ion-exchange columns and impractically large volumes of eluent.
Kanapilly and Newton (1971) have described a process applicable to the separation of even multi-Ci quantities of .sup.90 Y from .sup.90 Sr which overcomes many of the limitations of ion-exchange methods. This process, which involves the precipitation of .sup.90 Y as the phosphate, however, requires the addition of nonradioactive yttrium as a carrier, yielding a .sup.90 Y product which is not, obviously, carrier free. More recently, workers at Oak Ridge National Laboratory have introduced a procedure involving an initial solvent extraction of the .sup.90 Y from a dilute acid solution of .sup.90 Sr/.sup.90 Y using his (2-ethylhexyl) phosphoric acid in dodecane. Although good decontamination factors are routinely obtained, the useful life of the generator is limited, as radiolysis products of the extractant employed gradually accumulate in the .sup.90 Sr stock. Moreover, the complexity of the process, which involves repeated stripping of the initial extractant solution to reduce trace impurities and repeated wet ashing of the stock solution to destroy dissolved organic phosphates, is prohibitive.
U.S. Pat. 5,100,585, and U.S. application Ser. No. 08/076,881 filed Jun. 15, 1993, now U.S. Pat. No. 5,344,623 both assigned to the U.S. Department of Energy, and incorporated herein by reference, describe processes for the recovery of strontium and technetium values from acidic feed solutions containing other fission product values and containing up to 6 molar nitric acid. Recent work patented by the inventors, directed at improving analytical methodology for the determination of actinides and radiostrontium in biological and environmental samples, has led to the development of extraction chromatographic resins capable of selectively sorbing actinides and lanthanides or radiostrontium from nitric acid solution. See, Horwitz et. al, "Method for Liquid Chromatographic Extraction of Strontium from Acid Solutions", U.S. Pat. No. 5,110,474, and Horwitz et. al, "Method for the Concentration and Separation of Actinides from Biological and Environmental Samples", U.S. Pat. No. 4,835,107. The present application discloses the application of these materials to the preparation of .sup.90 Y of sufficient chemical and radiochemical purity to be suitable for use in medical applications.
Accordingly, it is an object of the present invention to provide a process for the preparation of .sup.90 Y of sufficient chemical and radiochemical purity to be suitable for use in medical applications.
It is another object of the present invention to provide a process for the separation of .sup.90 Y from .sup.90 Sr by selectively separating the .sup.90 Sr from the .sup.90 Y.
Yet another object of the present invention is to provide a process for the separation of .sup.90 Y from .sup.90 Sr which avoids the need to frequently replace the .sup.90 Sr and the generation of large quantities of radioactive waste.